球床式高温气冷堆球流混流的影响分析

研究球床式高温气冷堆球流存在的混流对堆芯关键参数的影响。开发了能模拟球流混流过程与效果的MFVSOP程序。选择球床模块式高温气冷堆核电站示范工程(HTR-PM)平衡堆芯为研究对象,对比分析不同的混流程度对堆芯功率峰值、功率密度等参数的影响及其不确定性。分析发现,混流对球床式高温气冷堆关键参数的不确定性影响不大,多次通过的燃料循环方式可降低不确定性。     

球床式高温气冷堆MOX燃料循环优化

与压水堆相比,球床式高温气冷堆能在堆芯结构不做明显改变的情况下采用全堆芯装载混合氧化物（MOX）燃料元件。基于250 MW球床模块式高温气冷堆堆芯结构,设计了4种球床式高温气冷堆下MOX燃料循环方式,包括铀钚混合的燃料球和独立的钚球与铀球混合装载的等效方式,采用高温气冷堆设计程序VSOP进行分析,比较了初装堆的有效增殖因数、燃料元件在堆芯内滞留时间、卸料燃耗、温度系数等主要物理特性。结果表明：采用纯铀和纯钚两种分离燃料球且铀燃料球循环时间更长的方案,平均卸料燃耗较高,总体性能较其他循环方式优越。     

250 MW球床模块式高温气冷堆进水事故研究

基于250 MW球床模块式高温气冷堆(HTR-PM)的初步设计,以高温气冷堆专用系统分析软件TINTE程序为主要工具,对蒸汽发生器1根传热管双端断裂设计基准的进水事故进行了分析,研究了反应堆温度和压力的变化特性、球床石墨的腐蚀率以及安全阀开启所造成的可燃气体排放等.此外,还分析了风机挡板关闭失效情况下堆内温度分布差异所造成的自然循环对事故后果的影响.计算结果表明:在蒸汽发生器1根传热管双端断裂、最大进水量600 kg情况下,事故后燃料元件的最高温度远低于设计限值,化学反应所引起的石墨腐蚀不会造成反应堆结构强度的破坏和燃料元件的意外破损,释放到反应堆舱室的可燃气体含量也不存在爆炸危险.     

球床式高温气冷堆的余热不确定性分析

反应堆在停堆后相当长时间内仍具有较高的剩余发热是核电站的重要特性,也是核电站安全分析的关键。因此,对反应堆余热及其不确定性进行分析,对于合理设计余热排出系统、研究论证燃料元件在事故后的安全特性等均具有重要意义。本工作结合德国针对球床式高温气冷堆制定的余热计算标准,介绍了球床式高温气冷堆剩余发热及其不确定性的计算方法,并结合200 MWe球床模块式高温气冷堆示范工程（HTR-PM）的初步物理设计,对长期运行在满功率平衡堆芯状态下的反应堆停堆后的余热及其不确定性进行了计算分析,为进一步的事故分析提供依据。     

高温气冷堆和轻水堆超铀元素管理特性研究

乏燃料中长寿命锕系元素对环境造成长期潜在危害,本文研究球床高温气冷堆不同燃料循环中超铀元素的产生和焚烧特性。在250 MW球床模块式高温气冷堆示范电站HTR-PM铀钚循环的乏燃料中提取铀和钚作为核燃料,设计了PuO₂和MOX燃料元件,将新设计的燃料元件重新装入与HTR-PM相同结构和尺寸的堆芯,分别形成纯钚燃料循环和MOX燃料循环。采用高温气冷堆物理设计程序VSOP,研究了高温气冷堆一次通过燃料循环和不同闭式燃料循环的超铀元素焚烧特性,并与轻水堆燃料循环结果进行比较和分析。结果表明：高温气冷堆一次通过燃料循环超铀元素生成率约为轻水堆的1/2;高温气冷堆闭式燃料循环能有效嬗变超铀元素。     

HTR—10石墨球与燃料球均匀混合装料初装堆方案研究

分析了球床式高温气冷堆几种可能的初装堆方案的特点，选取石墨球与燃料球均匀混合作为１０ＭＷ高温气冷实验堆的初装堆方案。利用高温气冷堆物理设计程序ＶＳＯＰ进行计算，分析屯ＨＴＲ－１０从初始装料向平衡态过渡过程中的倒换料方式，最大单球功率及最大燃耗变化情况。     

核级石墨辐照变形对辐照应力和寿命的影响

石墨是高温气冷堆的堆芯关键结构材料,其机械性能,尤其是辐照后特性,对反应堆的运行安全至关重要.不同牌号的石墨在制备工艺上有较大差异,导致内部微观结构的不同,从而影响石墨的辐照变形.本工作通过对高温气冷堆堆芯侧反射层石墨砖的辐照行为进行数值仿真,分析不同石墨材料的辐照变形对石墨结构的辐照应力和辐照寿命的影响.结果表明,石墨结构的辐照应力和辐照寿命对石墨材料的辐照变形高度敏感.相关结论将为高温气冷堆堆芯石墨砖的结构设计提供重要的数值依据.     

HTR-PM两根一回路连接管断裂的进气事故分析

进气事故是模块式高温气冷堆关注的超设计基准事故之一,石墨氧化腐蚀反应可能导致反射层结构强度减弱、燃料元件完整性和包容裂变产物能力被破坏,以及产生可燃气体等较严重后果。进气事故的分析研究对进一步掌握高温气冷堆的事故特性以及提高反应堆的安全设计具有重要意义。本文基于200MWe球床模块式高温气冷堆示范工程（HTR-PM）的初步设计,假设与一回路压力边界上、下相连的燃料元件进料管和卸料管同时发生断裂,从而形成烟囱效应并导致空气进入堆芯,利用高温气冷堆专用系统分析软件TINTE对自然循环建立及后续的进气腐蚀过程进行了研究,分析了自然循环流量、堆内石墨腐蚀速率、舱室氧气消耗量、燃料元件温度等关键参数的变化。结果表明,即使考虑腐蚀反应的不均匀性,事故后约60h时才会出现首个燃料包覆颗粒裸露现象,燃料元件最高温度峰值低于1620℃的设计限值,保持完好的燃料包覆颗粒仍具有包容放射性裂变产物的能力。同时,如果在相应的时间内采取措施切断进气源,使石墨腐蚀反应不能继续发展,将不会对反应堆的安全造成严重的影响。     

350MW环型球床高温气冷堆失冷失压事故下的热工特性研究

由于环型球床高温气冷堆特殊的堆芯结构，使其在失冷失压事故下堆内最高温度能够明显低于模块式球床高温气冷堆在相同事故下堆内最高温度。当堆芯热功率有较大幅度提高时，环型堆芯仍然能够凭借自身传热机能将衰变热量及时排出，满足失冷失压事故下燃料最高温度限制。这不仅增大了反应堆的安全性能，同时也能够有效地增加反应堆单堆功率，使环型球床高温气冷堆在经济上更具竞争力。本文研究环型球床高温气冷堆在提高功率水平时，反应堆在失冷失压事故下堆内的热工特性，并综合分析了几个重要的结构尺寸热工参数对失冷失压事故下燃料最高温度的影响。     

350兆瓦环形球床式高温气冷堆(英文)

提出了一个带环形球床堆芯高温气冷堆设计方案。它可以在保持模块式高温堆良好固有安全性的前提下,将模块式高温堆的功率从200MWt提高到350MWt。文章给出了主要事故安全的分析结果。为解决环形球床堆出口气体温差大的问题,提出了一种专门设计的堆体下部气流混合装置。混流装置模拟实验的结果表明,该结构能满足气流混合的要求。     

Evaluation of aseismic integrity in the HTTR core-bottom structure V. On the static and dynamic behavior of graphitic HTTR key-keyway structures

The graphite components in high temperature gas-cooled reactors are connected to each other through a key-keyway structure that has gaps between the key and the keyway to accomodate thermal expansion. Because a dynamic load concentrates on the key-keyway structure during earthquakes, it is considered to be a crucial element for assessing the integrity of the graphite components. A combination of experiments and analyses was employed to investigate the dynamic behavior of the key-keyway structure, i.e. the equivalent stiffness associated with vibrational characteristics of the graphite components and the stress distribution under dynamic loading. The experiments were performed using a graphite scale model and a dynamic photo-elastic method. The analysis was carried out using the finite element method (FEM) code Abaqus, taking account of the contact between the key and the keyway. The following conclusions were derived. (1) The equivalent stiffness of the key-keyway structure shows nonlinearity, owing to the contact deformation. (2) The equivalent stiffness evaluated by the FEM analysis, taking account of the non-inear contact deformation, is applicable for predicting the vibrational characteristics of ky-keyway structure. (3) The stress concentration under dynamic loading is lower than or nearly equal to that under static loading. The maximum stress concentration of the seismic load can be sufficiently evaluated under static loading conditions.     

V型凹口对液态熔盐堆用核石墨构件应力分析的影响

本文利用了一个根据球床模块堆(Pebble Bed Modular Reactor,PBMR)用核石墨材料辐照性能数据编写的用户自定义材料模型(User defined Material model,UMAT),按照美国橡树岭国家实验室(Oak Ridge National Laboratory,ORNL)的液态燃料熔盐试验堆(Molten Salt Reactor Experiment,MSRE)用核石墨构件尺寸,为钍基熔盐堆(Thorium-based Molten Salt Reactor,TMSR)设计了一款方型核石墨构件。利用新编UMAT对该核石墨构件进行了初步的应力分析。分析结果表明,在没有预制裂纹的情况下辐照梯度越大核石墨构件中心区域最大主应力值越大,构件的断裂位置可能出现在构件中心位置处;对于有V型凹口预制裂纹的情况,应力集中部位均出现在预制裂纹尖端附近,这将可能导致裂纹尖端附近出现裂纹扩展,从而引起构件断裂失效。     

Evaluation of aseismic integrity in the HTTR core-bottom structure IV. Structural integrity of connecting elements between graphite components

Former investigation on the aseismic test for core-bottom structure in the high temperature engineering test reactor (HTTR) clarified the response of acceleration, strain, impact load etc. Following this investigation the component test of connecting elements between graphite components was carried out to evaluate their fracture load and fracture mode. The stress analysis was also performed to estimate the stress profile by the analytical approach. The seismic design load is also estimated by considering the load concentration factor. In this paper the experimental and analytical results are compared and the structural integrity of these connecting elements is discussed under the severest earthquake condition postulated in the HTTR structural design.     

孔隙率对核级石墨氧化的影响

核级石墨在高温气冷堆中作为结构材料、慢化材料和反射层材料等被广泛应用,其氧化性能对高温气冷堆在进水或进气事故下材料的腐蚀行为有重要影响。初始孔隙率分布及孔隙率在氧化过程中的变化均对石墨氧化造成影响。本文以核级石墨IG-110、H-451、NBG-18和A3-3为例,以直径为6 cm的石墨球为研究对象,在一维瞬态氧化模型的基础上,分析了初始孔隙率分别服从均匀分布、正态分布和对数正态分布时对石墨氧化的影响。从模型简化和高温气冷堆安全分析角度保守考虑,建立石墨氧化模型时,核级石墨初始孔隙率可取均匀分布,此时石墨的整体失重率最大。     

Probabilistic failure analysis of a water-cooled tungsten divertor: Impact of embrittlement

Inherent brittleness and neutron embrittlement are critical weaknesses of tungsten for fusion application. Pronounced scattering of the fracture strength of tungsten requires a statistical treatment. Thus, the risk of structural failure of a tungsten component can be estimated only in a probabilistic framework. In this work, we applied a probabilistic failure analysis code STAU to estimate the failure risk of a water-cooled tungsten mono-block divertor component. The STAU code was based on the weakest-link failure theory and linear elastic fracture mechanics. A typical heat flux load being expected for a fusion reactor was considered for the FEM stress analysis. The failure probability was computed considering various mixed-mode fracture criteria. Both the experimentally estimated and hypothetical Weibull parameters were used as material data. In the case of unirradiated tungsten, the failure probability was acceptably small whereas reduced Weibull parameters led to significantly increased failure risk.     

高温气冷堆出口连接管束的流致振动研究

基于实验和数值模拟分析，以高温气冷堆蒸汽发生器传热管为对象，研究了流体攻角和截面圆角半径对带圆角的矩形截面柱（简称圆角方柱）旋涡脱落共振和驰振的影响，得到了不同攻角和不同圆角半径下圆角方柱的绕流特性，包括斯特劳哈尔数（St）和升阻力系数。研究结果表明，随着圆角半径的增大，结构的St增大，旋涡脱落共振对应的流速降低，同时驰振力系数逐渐增大，结构发生驰振的临界流速降低，但发生驰振失稳的攻角范围有所减小。本文研究为高温气冷堆蒸汽发生器传热管束矩形截面结构的流致振动设计提供了重要的依据。     

The application of the boundary-integral equation method to the solution of engineering stress analysis and fracture mechanics problems

This paper reviews the development and application of an influence function method for calculating stress intensity factors and residual fatigue life for two- and three-dimensional structures with complex stress fields and geometries. Through elastic superposition the method properly accounts for redistribution of stress as the crack grows through the structure. The analytical methods utilized and the computer programs necessary for computation and application of load independent influence functions are presented. A new exact solution is obtained for the buried elliptical crack, under an arbitrary Mode I stress field, for stress intensity factors at four positions around the crack front. The IF method is then applied to two fracture mechanics problems with complex stress fields and geometries. These problems are of current interest to the electric power generating industry and include (1) the fatigue analysis of a crack in a pipe weld under nomial and residual stresses and (2) fatigue analysis of a reactor pressure vessel nozzle corner crack under a complex bivariate stress field.     

Failure strains and proposed limit strains for an reactor pressure vessel under severe accident conditions

The local failure strains of essential design elements of a reactor vessel are investigated. The size influence of the structure is of special interest. Typical severe accident conditions including elevated temperatures and dynamic loads are considered.The main part of work consists of test families with specimens under uniaxial and biaxial load. Within one test family the specimen geometry and the load conditions are similar, but the size is varied up to reactor dimensions. Special attention is given to geometries with a hole or a notch causing non-uniform stress and strain distributions typical for the reactor vessel. A key problem is to determine the local failure strain. Here suitable methods had to be developed including the so-called “vanishing gap method”, and the “forging die method”. They are based on post-test geometrical measurements of the fracture surfaces and reconstructions of the related strain fields using finite element models.The results indicate that stresses versus dimensionless deformations are approximately size independent up to failure for specimens of similar geometry under similar load conditions. Local failure strains could be determined. The values are rather high and size dependent. Statistical evaluation allow the proposal of limit strains which are also size dependent. If these limit strains are not exceeded, the structures will not fracture.     

Component exposure in hypothetical accidents with very fast depressurization in a HTR module reactor

The starting event of the massive air ingress into the core of the HTR module reactor, classified as hypothetical incident, is the very fast depressurization of the primary circuit. Provided that the integrity of the reactor pressure vessel is not in question, a rupture of the connecting pressure vessel between reactor pressure vessel and steam generator vessel is the maximum possible leak cross-section. In this work it is investigated whether the components of the reactor pressure vessel are exposed by the depressurization process to mechanical loads which exceed the load limits. These loads are caused by two different events, the strong momentum change of the fluid and the local pressure differences, respectively. Due to the momentum change the bottom reflector receives the maximum load, whereby only 2% of the compressive strength of the graphite quality used there are reached. However, the load by local pressure differences is between passed volumes and in normal operation, not-passed volumes lead to high load values. A maximum pressure difference of 44.5 bar was calculated at the thermal top shield.     

高温翅片管的长时蠕变屈曲分析及评定方法研究

由于翅片管处于高温环境和外压载荷下,需要考虑其发生蠕变屈曲失效的风险。本文对翅片管在高温环境下的蠕变屈曲分析及评定方法进行了研究,提出了一种基于塑性本构和蠕变本构的有限元长时蠕变屈曲分析方法,并通过数值拟合,获得了高温屈曲的失效评定图以及失效评定公式,提出了一种方便应用于工程的快速评定方法。针对翅片管结构,将该方法的评定结果与规范中的屈曲分析评定结果进行对比,验证了该方法的可行性。同时研究了在有压力波动的情况下,结构的临界屈曲时间与载荷历程的关系,为复杂结构和复杂载荷工况的蠕变屈曲分析奠定了基础。